openmsr / msr-archive
curated documents pertaining to early molten salt reactor research.
☆21Updated 8 months ago
Related projects ⓘ
Alternatives and complementary repositories for msr-archive
- List of open source projects related to OpenMC☆20Updated 2 months ago
- Native plotting GUI for model design and verification☆47Updated 2 weeks ago
- Stochastic Calculator Of Neutron transport Equation☆35Updated last month
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆52Updated 4 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆12Updated 5 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆24Updated 2 months ago
- Tool for converting MCNP input files to OpenMC classes/XML☆36Updated 2 weeks ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆32Updated 3 years ago
- Catalogs a plethora of reactor inputs for OpenMC☆18Updated 3 years ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆57Updated 2 weeks ago
- ☆17Updated 2 months ago
- Python ENDF Parser☆26Updated 3 months ago
- ☆28Updated last year
- OpenMC wrapped as a Moose App☆17Updated last year
- Tools to translate between different CSG geometry types☆35Updated last year
- Sampling nuclear data and uncertainty☆48Updated last month
- Collection of models for reactor physics benchmarks☆48Updated last month
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆36Updated 4 months ago
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆34Updated last week
- Collection of scripts for managing data for OpenMC☆27Updated 11 months ago
- open-source Sn software☆21Updated this week
- OpenMC Monte Carlo Code☆14Updated 2 months ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆63Updated last year
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆17Updated last year
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆32Updated this week
- Nuclear data processing with legacy NJOY☆98Updated 2 weeks ago
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago